Reactor Outlet Coolant Temperature Research Articles

Dynamic characteristics of a recompression supercritical CO2 cycle against variable operating conditions and temperature fluctuations of reactor outlet coolant

The transient performance of an inventory-controlled supercritical CO2 recompression cycle under synchronous adjustment of parameters has rarely been studied. In this study, a one-dimensional design of a supercritical CO2 recompression cycle with a 10 MW output power was completed for a lead-cooled fast reactor. The dynamic characteristics of the system power, which decreased from full load to half load under three ramp signals, were compared. Furthermore, the effect of the fluctuating reactor outlet coolant temperature on the transient performance of the system was investigated. Steady-state analysis revealed that the maximum design efficiency of the system was 43.40% at a split ratio of 0.37. The dynamic results showed that the shorter the ramp signal for parameter adjustment, the more significantly the transient performance of the system degraded. When the load decreased significantly, a secondary efficiency trough appeared in the system. These poor phenomena were improved by appropriately extending the action time of the ramp signal. Additionally, the influence of coolant temperature fluctuation on the temperature at both ends of the recuperator differed owing to the difference in the heat capacity. Moreover, the reduction in the target load of the system made the fluctuation range wider for efficiency but narrower for power.

Fuzzy logic-based power control system of a supercritical-pressure light water fast reactor

Supercritical-pressure light water fast reactor is one of the current concepts of the Generation IV reactors which have significant improvement on the efficiency, safety and economy. A change of power to flow ratio leads to a change of the outlet coolant temperature in second. The temperature deviation must be suppressed as small as possible to keep the integrity of the reactor structures, like the reactor outlet nozzles, main steam line, etc. The plant dynamics analysis was carried out by giving perturbations of power to flow ratio, and then its influence on the reactor power and outlet coolant temperature were observed. A fuzzy logic-based power control system was designed and applied to the plant to compensate the effects of the perturbations. The control’s parameters, like the member functions of fuzzification were optimized to satisfy the criteria of allowable outlet coolant temperature deviation and stability. The power could be controlled stably by the optimized control design and the criterion of the outlet coolant temperature deviation is satisfied against perturbations of power to flow ratio.

Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 30% reactor power

In a safety demonstration test involving the loss of both reactor-reactivity control and core cooling, the High-Temperature engineering Test Reactor (HTTR) demonstrated spontaneous stabilization of the reactor power. A test where all three gas circulators were tripped at 30% reactor power (9MW) without a reactor scram, called a loss-of-forced-cooling (LOFC) test, was analyzed, and the analytical results from a reactor kinetics code was reported. After that, the Japan Atomic Energy Agency (JAEA) acquired a large supercomputer and the STAR-CCM+ software developed by CD-adapco, which can be used to create a three-dimensional model of the HTTR to analyze steady-state and transient conditions. In this paper, a model of the three-dimensional thermal-hydraulics inside the reactor pressure vessel (RPV) during the LOFC test at 30% reactor power (9MW) is described and detailed analyses are performed using STAR-CCM+. With this tool, the effects of natural and forced convection on thermal-phenomena in the HTTR can be understood quantitatively. The finding is as follows: the downstream of forced convection caused by the helium purification system (HPS) pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection due to the HPS has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. And the effect can be neglected because the difference of the maximum velocity of the helium between the cases where the HPS is active and inactive, is very little over the course of LOFC test. The proposed three-dimensional analytical model will be useful for three-dimensional accident analyses.

Dynamic Modeling and Control Characteristics of the Two-Modular HTR-PM Nuclear Plant

The modular high temperature gas-cooled reactor (MHTGR) is a typical small modular reactor (SMR) with inherent safety feature. Due to its high reactor outlet coolant temperature, the MHTGR can be applied not only for electricity production but also as a heat source for industrial complexes. Through multimodular scheme, that is, the superheated steam flows produced by multiple MHTGR-based nuclear supplying system (NSSS) modules combined together to drive a common thermal load, the inherent safety feature of MHTGR is applicable to large-scale nuclear plants at any desired power ratings. Since the plant power control technique of traditional single-modular nuclear plants cannot be directly applied to the multimodular plants, it is necessary to develop the power control method of multimodular plants, where dynamical modeling, control design, and performance verification are three main aspects of developing plant control method. In this paper, the study in the power control for two-modular HTR-PM plant is summarized, and the verification results based on numerical simulation are given. The simulation results in the cases of plant power step and ramp show that the plant control characteristics are satisfactory.

A rapid evaluation method of the heat removed by a VCS before rise-to-power tests

Before rise-to-power tests, the actual measured value of heat released from the Reactor Pressure Vessel (RPV) or removed by the Vessel Cooling System (VCS) cannot be obtained. It is difficult for operators to evaluate the reactor outlet coolant temperature supplied from the High Temperature Engineering Test Reactor (HTTR) before rise-to-power tests. Therefore, when the actual measured value of heat released from the RPV or removed by the VCS are changed during rise-to-power tests, operators need to evaluate quickly, within a few minutes, the heat removed by the VCS and the reactor outlet coolant temperature of 30 MW, at 100% reactor power, before the temperature achieves 967°C which is the maximum temperature limit generating the reactor scram. In this paper, a rapid evaluation method for use by operators is presented. The difference between the experimental and analytical results was within 30 (kW) and it is appropriate that the presented evaluation method can be applied; therefore, operators can analyze the heat removed by the VCS quickly, within a few minutes, before and during Rise-to-Power Tests.

GTHTR300—A nuclear power plant design with 50% generating efficiency

GTHTR300 is a gas turbine high temperature reactor power generation plant design. The baseline design reported by Japan Atomic Energy Agency a decade ago attained 45.6% net efficiency. Technological improvements have since been made that make further increase in efficiency practical: first, the cycle parameters are upgraded by utilizing the newly acquired design data including those from component tests. Next, the core design is optimized to raise the reactor outlet coolant temperature from the baseline of 850°C to the level of 950°C demonstrated on the long-term test reactor operation. Both core physics and thermal hydraulics are investigated to demonstrate the corresponding temperature rise is within the design limit so that the existing fuel design can continue to apply. Finally, an advanced type of turbine blade material that has only recently entered in commercial service in aircraft engine is found to be useable for this design to realize a turbine inlet temperature of 950°C without requiring blade cooling.As detailed in this paper, these design improvements result in a nearly 5% gain in overall plant efficiency and enable the GTHTR300 to break the 50% efficiency barrier of nuclear plant while using only the existing technologies. This result is expected to contribute to the early market deployment of high temperature gas-cooled reactor.

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30MW and a maximum reactor outlet coolant temperature of 950°C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9MW to 0MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the strength of the neutron source, the larger the minimum reactor power.

A Small-Sized HTGR System Design for Multiple Heat Applications for Developing Countries

Japan Atomic Energy Agency has conducted a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for multiple heat applications, named HTR50S, with the reactor outlet coolant temperature of 750°C and 900°C. It is first-of-a-kind of the commercial plant or a demonstration plant of a small-sized HTGR system to be deployed in developing countries in the 2020s. The design concept of HTR50S is to satisfy the user requirements for multipurpose heat applications such as the district heating and process heat supply based on the steam turbine system and the demonstration of the power generation by helium gas turbine and the hydrogen production using the water splitting iodine-sulfur process, to upgrade its performance compared to that of HTTR without significant R&D utilizing the knowledge obtained by the HTTR design and operation, and to fulfill the high level of safety by utilizing the inherent features of HTGR and a passive decay heat removal system. The evaluation of technical feasibility shows that all design targets were satisfied by the design of each system and the preliminary safety analysis. This paper describes the conceptual design and the preliminary safety analysis of HTR50S.

Thermal Performance of Intermediate Heat Exchanger during High-Temperature Continuous Operation in HTTR

To use the High-Temperature Gas-cooled Reactor (HTGR) system as a commercial reactor similar to the future HTGR ‘GTHTR300C,’ supplying stable high-temperature heat from the reactor to the heat utilization system such as the hydrogen production system over a prolonged period should be demonstrated. The 50-day high-temperature continuous operation with full power, which is the first long-term operation with a reactor outlet coolant temperature over 900°C, was achieved in March 2010 in the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA). In order to supply stable high-temperature heat from the reactor to the heat utilization system using the HTTR, the heat exchange performance and structure integrity of the intermediate heat exchanger (IHX) should be confirmed. In this paper, the heat exchange performance and structure temperature of the IHX were evaluated with the high-temperature continuous operation data and compared with the designed one or estimated one. As a result, it was confirmed that the IHX can exchange stable high-temperature heat from the primary coolant to the secondary helium, and the IHX structure temperature is under the allowable working temperature in the operation. Moreover, it was confirmed that the heat exchange performance is almost the same as the designed one and that the structure temperature estimated with the analysis code is almost the same as the measured one.

ICONE19-43224 ANALYSIS OF A LOSS OF FORCED COOLING TEST USING THE HIGH TEMPERATURE ENGINEERING TEST REACTOR (HTTR)

The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950℃ (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and xenon reactivities) affecting re-critical time and reactor peak power level and total reactivity is addressed. The analytical results will be utilized for the design and construction of the Kazakhstan High Temperature Reactor (KHTR) and the realization of commercial Very High Temperature Reactor (VHTR) systems.

ICONE15-10158 CORE DYNAMICS ANALYSIS FOR REACTIVITY INSERTION AND LOSS OF COOLANT FLOW TESTS USING THE HTTR

The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950℃ (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed. The reactor power decreases to decay heat level from the maximum reactor power of 30 MW due to the negative reactivity feedback effect of the core. Although the reactor power becomes critical again about five hours later, the peak power value is merely 2 MW. It was confirmed that by using the developed code, it is possible to not only analyze the reactor core dynamics but also simulate the core dynamics during the abnormal events postulated in the HTGR safety analysis.

ICONE15-10222 IMPROVEMENT OF ANALYSIS TECHNOLOGIES FOR HTGR BY USING THE HTTR DATA

The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950℃ in April, 2004 during the "rise-to-power tests" confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with the high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.

ICONE15-10205 VALIDATION OF NEUTRONICS CALCULATION CODES FOR VHTR NUCLEAR DESIGN USING HTTR EXPERIMENTAL DATA

In order to validate applicability of neutronics calculation codes for a nuclear design of the VHTR, core calculations have been performed for the HTTR. Additionally, effects of difference of nuclear data libraries on the core calculations for the HTTR have been studied for JENDL(Japan), ENDF/B(U.S.A.) and JEFF(Europe). The HTTR is a graphite-moderated and a helium gas-cooled reactor constructed at JAEA O-arai Research and Development Center. The first criticality was achieved in 1998. A rated thermal power of 30MW and the reactor-outlet coolant temperature of 950℃ were achieved in 2004. Three-dimensional core calculations for the HTTR excess reactivity at room temperature condition were performed by different two methods. One was a based on a diffusion theory using the SRAC code system and another was a based on a Mote-Carlo theory using the MVP code. The SRAC will be used mainly for the VHTR nuclear design in JAEA and the MVP will be used as the reference. In the calculations, both codes were used with the neutron cross-section sets generated from JENDL-3.3 which is the latest version of the JENDL, respectively. A heterogeneous effect caused by coated fuel particles on the calculations was taken into consideration by using the functions in these codes. In the SRAC calculations, a suitable lattice model using the lattice calculation was discussed from the point of calculated neutron energy spectrum. In consequence, the calculation result of the HTTR excess reactivity at room temperature condition by the MVP was in good agreement with the experimental data within 0.4%Δk/k and that by the SRAC, meanwhile, overestimated the experimental data about 1.5%Δk/k. Calculations for the HTTR excess reactivity were performed using the MVP with three different nuclear data libraries JENDL-3.3, ENDF/B-VI.8 and JEFF-3.1, respectively. In consequence of the comparison between these calculation results and the experimental data, JENDL-3.3, ENDF/B-VI.8 and JEFF-3.1 yielded the excess reactivity agreement with the experiments within 0.4%Δk/k, 0.7%Δk/k and 0.7%Δk/k, respectively. The discrepancy of the excess reactivity between ENDF/B-VI.8 and JEFF-3.1 was less than the standard deviation and thus it was negligible. Furthermore, identifications of nuclides which make major contributions to the discrepancy of the excess reactivity was studied. As the results, the discrepancy between JENDL-3.3 and ENDF/B-VI.8 was mainly caused by the difference of graphite data.. In order to treat S(α,β) data, the MVP neutron cross-section set generated from JENDL-3.3 was taken it from ENDF/B-VI, so that S(α,β) data for graphite of JENDL-3.3 is same as in ENDF/B-VI.8. Therefore, the discrepancy of excess reactivity between JENDL-3.3 and ENDF/B-VI.8 were mainly caused by difference of carbon free-gas data.

Achievement of Reactor-Outlet Coolant Temperature of 950°C in HTTR

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30 MW and reactor-outlet coolant temperature of 950°C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950°C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950°C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.

Overview of HTTR design features

The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850 °C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by 2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.

Performance test of HTTR

The high temperature gas-cooled reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The high temperature engineering test reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 850 °C on 7 December 2001 during the ‘rise-to-power tests’. Two kinds of tests were carried out during the ‘rise-to-power tests’. One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, and control system. From the test results of the ‘rise-to-power tests’ up to 30 MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility could be performed safely.

Design of core components

The high temperature engineering test reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with a thermal output of 30 MW and a reactor outlet coolant temperature of 950 °C during the high temperature test operation. Fuel elements, control rod guide blocks and replaceable reflector blocks are main core components of the HTTR. The core components are mainly made of ceramics, such as graphite, silicon carbide, etc. which are excellent thermal resistant materials. Design requirements for the core components were laid down and extensive design efforts and R&D were performed in order to maintain their structural integrity. The structural integrity of the core components was ascertained by performing evaluations on the basis of the R&D and stress analysis results. This paper describes the design requirements and design details of the core components, and the structural integrity evaluation results for the fuel and graphite blocks.

Achievement of Reactor-Outlet Coolant Temperature of 950.DEG.C. in HTTR

Achievement of Reactor-Outlet Coolant Temperature of 950.DEG.C. in HTTR

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 °C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open.